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    STP1295

    Comparison of the Long-Time Corrosion Behavior of Certain Zr Alloys in PWR, BWR, and Laboratory Tests

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    Abstract

    Laboratory corrosion tests have always been an important tool for Zr alloy development and optimization. However, it must be known whether a test is representative for the application in-reactor. To shed more light on this question, coupons of several Zr alloys were exposed under isothermal conditions in all or most of the following environments: In-Reactor: (1) PWR core at 300 to 340°C up to six years. (2) BWR core with a low sensitivity to nodular corrosion up to four years. (3) BWR core with a high sensitivity to nodular corrosion up to two years.

    Ex-Reactor (in Autoclave): (1) 350°C/pressurized water up to three years. (2) 400°C/100-bar steam up to two years. (3) 350°C/0.01 M LiOH water up to two years. (4) 500 to 515°C/high-pressure steam 16 to 24 h.

    In addition, the material condition of several of the examined Zr alloys was varied over a wide range.

    For evaluation of the in-PWR tests and for comparison of out-of-pile and in-pile tests, the different temperatures and times were normalized to a temperature-independent “normalized time” by assuming an activation temperature (Q/R) of 14 200 K. Comparison of in-PWR and out-of-pile corrosion behavior of Zircaloy shows that corrosion deviates to higher values in PWR if a weight gain of about 50 mg/dm2 is exceeded. In the case of the Zr2.5Nb alloy, a slight deviation of corrosion as compared to laboratory results starts in PWR only above a weight gain of 100 mg/dm2. In BWR, corrosion of Zircaloy is enhanced early in time if compared with out-of-pile. Zr2.5Nb exhibits higher corrosion results in BWR than Zircaloy-4.

    Alloying chemistry and material condition affect corrosion of Zr alloys. However, several of the material parameters have shown a different ranking in the different environments. Nevertheless, several material parameters influencing in-reactor corrosion like the second phase particle (SPP) size or in-PWR behavior as the Sn and Fe content can be optimized by out-of-pile corrosion tests.

    Keywords:

    zirconium alloys, Zircaloy, corrosion, in-PWR corrosion, in-BWR corrosion, out-of-pile corrosion, neutron irradiation, radiation effects, nuclear application


    Author Information:

    Garzarolli, F
    Siemens AG, Power Generation Group (KWU), Erlangen,

    Broy, Y
    Siemens AG, Power Generation Group (KWU), Erlangen,

    Busch, RA
    Siemens Power Corporation, Richland, Washington


    Committee/Subcommittee: B10.01

    DOI: 10.1520/STP16204S